Boiling Water Reactors are large machines designed for electric power generation using nuclear fission as a heat source. Power is generated in the reactor core which is placed inside a large pressure vessel. The reactor core is made up of an arrangement of fuel assemblies also called fuel bundles. Typically, there are 400˜800 fuel assemblies in a BWR core. Each of the fuel assemblies is arranged inside a vertical channel of square cross section through which water coolant is injected from the bottom. Each of the fuel assemblies consist of a plurality of vertical rods arrayed within the said vertical channels in a typically 7×7, 8×8, 9×9, or 10×10 rod matrix. The said rods are sealed cylindrical tubes inside which ceramic pellets of fissionable material, e.g. Uranium oxide, are stacked. The fuel rod tubes, also called cladding, and the outer channel encasing each fuel assembly, are made of a low neutron absorbing metal such as Zirconium-based alloys. The fuel rods are spaced and supported by an upper a lower tie plates. Mechanical support to maintain the rod positions against possible bow under thermal or mechanical loads is provided by several spacer grids. A handle attached to the upper tie plate is used for transporting the fuel assembly. The lower tie plate is the coolant flow entry point and is designed to fit the core support plate.
The water flows upward in the fuel channels and removes the heat generated in the pellets by the fission of the uranium and plutonium nuclei. In addition to its cooling function, the water serves as neutron moderator. The neutron moderation function is achieved as the neutrons produced in the fission process collide with the hydrogen atoms in the water molecules and slow down to lower energies which increase the probability of inducing further fission reactions.
In Boiling Water Reactors, the water coolant is allowed to boil as it travels up in each fuel assembly channel. The density of the coolant is reduced by the boiling process and consequently the moderating function is adversely affected particularly in the upper part of the fuel assembly where the fuel-to-moderator ratio becomes higher than optimally desired. This problem was mitigated in some fuel assembly designs by introducing one or more water rods or channels, henceforth called water channels. A water channel is a hollow tube or conduit extending vertically along the fuel rods, and through which part of the water flows without boiling. Thus, the amount of water available for the neutron moderating function is increased. The said improvement in the moderation function comes at the expense of reducing the amount of water available for the cooling function. Another common improvement in the design fuel assemblies is the use of part-length fuel rods. While the typical active length of a full-length fuel rod is 3.8 m, few short rods in selected array positions are used. The length of a part-length rod is typically half to two-thirds that of the full length rod, and there are typically 8 to 12 part-length rods in each assembly. The space vacated by cutting down the length of some rods is filled with voided coolant (steam-water mixture) flow, and therefore restores the fuel-to-moderator ratio in the top part of the fuel assembly closer to the optimum value for nuclear criticality. The use of part-length fuel rods is also beneficial in reducing the flow resistance in the top part of the assembly as the flow area is increased. However, the use of part-length rods comes at the expense of the amount of fissionable material that can be packed into a fuel assembly.
The reactor core therefore is made of a number of parallel, nuclear-heated, boiling channels. The core is supported at the bottom with the so-called core support plate, where each of the fuel assembly is seated on a flow opening called inlet orifice. The inlet orifices restrict the flow into each fuel assembly and serve to distribute the total flow into the core evenly among individual fuel assemblies. The core is encased in a cylindrical shroud which separates the upward boiling flow inside the core from the downward flow in the downcomer, where the latter is the annulus space between the core shroud and the pressure vessel wall. The core shroud is capped at the top by a dome-like structure to form the so-called upper plenum. The liquid water and steam mixture flowing from the exit of the core fuel assemblies mix freely in the upper plenum and continue their upward flow into a set of parallel tubes called standpipes emanating from the upper plenum dome. Each standpipe is fitted with a steam separator device which directs almost dry steam into the upper part of the pressure vessel where it flows into the steam lines leaving the pressure vessels in order to drive steam turbines for the purpose of generating electric power. The saturated water leaving the steam separators is directed to flow down into the water pool that surrounds the standpipes and mix with the lower temperature feedwater returning from the condenser. This subcooled water flows downward in the downcomer being driven by a combination of the density head and that of an array of pumps placed in the downcomer. The water leaving the downcomer gather in the so-called lower plenum before it is distributed through the orifices at the bottom of the core, continue to flow upward and boil in the fuel channels, and exit the core thus completing the recirculation loop. In the special case of recirculation pumps not operating or tripped, the coolant flow occurs by natural circulation, that is driven solely due to density difference between the single phase side outside the core, and the two-phase side in the core and the riser assembly which consists of the upper plenum, the standpipes, and steam separators. Natural circulation is an operating state where the flow is reduced and the power-to-flow ratio is higher than the case with the pumps operating, thus prone to density wave instabilities.
The nuclear reaction is controlled by the so-called control rods which are neutron absorbing devices that can be moveable in the spaces between fuel assemblies and are driven by mechanisms under the core support plate thus occupying part of the space of the pressure vessel lower plenum. The control rods can be inserted quickly to shut down the chain reaction in the core, an operation called scram. The scram function is needed for terminating plant transients that have the potential of bringing the operation outside the plant design limits. Reactor scram can be initiated either automatically or by operator action. In some accident scenarios, the scram function is assumed to fail and the consequences of such hypothetical scenarios are used for risk assessment guiding the design and licensing of boiling water reactors.
Detailed description of BWR design and operation can be found in Ref. (1).
A central part of the background of this invention is the coolant flow and core power instabilities which must be avoided for safe and smooth operation of the BWR plant. The instability phenomena and the effects of the resulting power and flow oscillations are described below.
The reactor operation is stable under normal operating conditions, but can depart from stable configuration at conditions of typically high power combined with low flow. The nature of the instability is outlined below.
The unstable behavior in a BWR is associated with the density waves in vertical boiling channels such as BWR fuel assemblies. In the case of a random perturbation to the flow rate at the inlet of the channel, while the energy transfer rate to the coolant remains unchanged, a corresponding enthalpy wave travels upward with the flow.
Downstream from the elevation of boiling inception, the flow enthalpy is translated to a steam quality wave where more steam is generated per unit of flow rate to account for an enthalpy increase. The void fraction (by volume), defined as the ratio of the steam volume to the total volume, is generally proportional to the steam quality, and therefore a void fraction wave traveling up the boiling channel results from the originating inlet flow perturbation. The void fraction can be expressed in terms of the average flow density, where maximum density is associated with zero void content, and minimum density is associated with a void fraction of unity. We can therefore speak of a density wave which results from an originating inlet flow perturbation. All flow parameters, mainly flow rate and steam quality and void fraction, are subsequently perturbed and the perturbations travel upward in the boiling channel with a phase lag.
The density wave alters the flow characteristics in two ways. The first one is that the total weight of the coolant in the channel, which is proportional to the integrated density along the channel, is altered dynamically resulting in a net gravitational pressure head response. The second way is the change in friction pressure drop along the channel. The friction pressure drop in turn is affected in two ways: the first way is through the change in the flow rate itself (friction being proportional to the square of flow rate), and the second way through the change in the so-called two-phase multiplier which accounts for the increase in frictional pressure drop for higher steam quality. In an idealized situation, the net pressure drop across the channel is kept constant, which leaves a residual component of force to compensate for the driving changes in density head and the changes in friction. The net force accelerates the flow, which reinforces an original flow perturbation of the so-called resonant frequency leading to the potential growth of the oscillation. The density wave degree of stability is reduced for higher power-to-flow ratios and for bottom-peaked axial power distribution as they tend to increase the void content and subsequently the density head feedback which drives the instability. High friction pressure drop at the channel inlet increases kinetic energy dissipation and helps to stabilize density waves, while high friction at higher elevations is destabilizing due to the phase lag of their effect which tends to reinforce the original perturbation.
In a BWR, the oscillation of flow rate in the fuel bundles resulting from density waves is complicated by the double role the water plays in the operation of the reactor. The density wave results in a corresponding neutron moderation effectiveness which in turn results in nuclear reactivity and fission power responses. The fission energy generated inside the fuel rods is transferred to the coolant through heat conduction in the fuel rods through the clad surface. The fluctuation of the heat flux through the clad surface is filtered through the heat conduction processes through the fuel rods and the clad surface heat flux experiences a damped and delayed response relative to the fission power itself. The fluctuation of the heat flux results in corresponding fluctuations in the boiling rate and the coolant density where such feedback tends to further destabilize the density waves in the boiling channels.
The operation of BWR under oscillating conditions is not permitted by the Nuclear Regulatory Commission (NRC) in the US or its equivalent authorities in foreign countries. This restriction is placed in order to avoid violating the thermal limits in the fuel, potentially resulting in fuel damage.
A detailed report on density wave instabilities and oscillations in BWR's can be found in Ref. (2).
The prior art dealt with stability issues in various ways. In one way, new fuel designs aim at maintaining the level of stability as the preceding designs or better, but actual improvements could hardly be achieved without negatively impacting other parameters important to the economic performance of fuel designs such as power density. Modern fuel designs tend to include larger number of smaller diameter rods compared with older designs, which are less stable due to decreasing the rod heat conduction time constant. The use of part-length rods tends to stabilize the hydraulic flow by reducing flow resistance in the top part of the channel, but comes at the expense of reducing the mass of the fissionable material load in each fuel bundle. The use of water channels improves stability through reducing the relative dependence on the steam-water mixture coolant for neutron moderation, but it comes at the expense of reduced number of fuel rods. In general, fuel design modifications are not sufficient to achieve unconditional stability. Another way of dealing with BWR stability is limiting the degree of axial and radial power peaking variations anticipated in the design of a reload fuel cycle, which adversely affects the net energy that can be generated by the reactor in any given fueling cycle. The most effective way to deal with the potential for instability in the prior art is the operations option. In one of these operational solutions, the operation of the reactor is restricted inside a pre-calculated so-called exclusion zone, which is an area in the power-flow map characterized by high power-to-flow ratio. The other operational solution is the so-called detect and suppress (D&S) solution, where an automatic shut down is initiated upon detection of oscillatory behavior. In essence, protection of the reactor fuel against the consequences of power and flow instabilities is dependent on the ability to scram before the unstable oscillations grow to a magnitude that is sufficiently large to cause fuel damage.
A central part in defining the need for the present invention is the rare hypothetical event called Anticipated Transient Without Scram (ATWS) which potentially results in large unstable power and flow oscillations. Basically, an anticipated transient such as a turbine trip results in a sequence of control actions and the plant responds as designed by tripping the recirculation pumps bringing the core flow to the lower rate associated with natural circulation. The tripping of the turbine deprives the feed-water heaters from their steam supply and the temperature of the coolant entering the core is reduced. This combination of conditions, relatively high power and low coolant flow rate that enters the core at lower than normal operation temperature, is conducive to instability. The reactor protection system is designed to scram, however the adverse condition of failure to scram that is the so-called ATWS condition may lead to severe power and flow oscillations that have the potential of causing serious fuel damage. Simulations of such ATWS transients can be found in References [3] and [4]. It is shown in Reference [4] that the flow oscillations are so severe that the inlet flow rate in some fuel bundles reverses direction momentarily, that is the flow oscillation magnitude significantly exceeds the average flow rate; and the resulting calculated temperature of the fuel rod surface is sufficiently high to breach the cladding integrity.
The prior art recognizes the problem of unstable power and flow oscillations during ATWS, and mitigation measures are prescribed. As shown in Reference [3], the oscillations are damped and the reactor is successfully shut down upon the injection of boron in the reactor vessel. However, the boron injection cannot be accomplished without delay and severe oscillations can be reached in the meantime. Also, the subsequent cleanup of the reactor vessel to remove the injected boron before operation is resumed is difficult and expensive. The prior art is silent concerning any hardware or fuel design modifications that protect the reactor core from severe power and flow oscillations of an ATWS. The present invention provides this protection as described in the following sections.
The prior art addresses the need for trapping debris that may be carried by the cooling water through the lower tie plate where they come into contact with the fuel rods and could cause fretting of the cladding material and eventually fuel rods leaking. For this purpose, filters are installed in the lower tie plate. The present invention provides a novel means for stopping flow-carried debris, which can be relied on solely or in addition to a filter of a prior art design.
The problems of loss-of-coolant accidents are well known, which are addressed by injecting emergency core cooling water. Improving the efficiency of emergency core cooling is a standing problem in the nuclear technology. However, no hardware modifications could be found in the prior art to address this problem.